The present invention of a nuclear power fuel element provides for significant performance improvements of light waterreactor (LWR) fuel under heavy duty operating conditions e.g. under load follow, off-normal power transients, and extended fuel burnup operation, where phenomena like fission gas release and stress corrosion cracking of the fuel cladding may be life limiting. The invention resides in a significant improvement of the basic fuel design concept for use in current type nuclear fuel elements (fuel rods) in water cooled power reactors as described in British Pat. No. 1 454 618.
The characteristic design feature of said prior art is the fuel clad tubing (made out of zirconium metal or zirconium alloy), which is provided on its inside bore surface with a multitude of radially protruding, axially extending ribs of minute radial dimensions (&lt;1% of the tube radius). Said ribbed bore configuration is claimed to offer two phenomenologically different, but concurrent, fuel performance improvements:
1. It minimizes the propensity for fuel failure through pellet/clad interaction (PCI) induced stress corrosion cracking (SCC) on fuel power ramping during reactor operation by reducing the magnitude of stress concentrations imposed by widening crack openings in the underlying fuel pellets. PA1 2. It prevents the formation of local axial gaps in the fuel column under power operation due to in-reactor pellet densification (a safety concern) by locking the individual pellets in their axial positions at their slightly outbending and indenting corners (see FIGS. 6 and 7 of said British patent). PA1 1. The degradation of the pellet/gap thermal conductivity following FGR. PA1 2. The thermal feed back effect. PA1 3. The slow axial gas migration. PA1 4. The PCI/SCC failure propensity. PA1 5. The shorter fuel thermal "constant" of small diameter rods. PA1 According to the invention provisions are made in designing the clad bore configuration and the pellet shape in such a manner, that during power operation of the fuel rod under close interfacial mechanical contact between the individual chamfered fuel pellets and the ribbed clad tubing, a system of voids or channels of minute dimensions for gas communication remains open and interconnected in the pellet/clad interface all along the fuel column, and the element of this invention thus characteristically comprises: PA1 1. A set of axially extending narrow channels formed between the ribbed bore surface and the pellet column, and PA1 2. pellet-pellet interface volumes composed of (a) a set of circumferential ring-shaped, spacious channels formed between the chamfered pellet end corners and the clad tubing, and (b) any dish volumes at the pellet end surfaces. The channel system thus forms a regular grid structure composed of a multitude of axially oriented channels cross linked by a multitude or ring (torus) shaped channels.
The densification problem of the early seventies is, however, no longer of any licencing concern to the nuclear fuel industry. Already by the mid seventies that problem was resolved by other means, such as by proper control of the pellet sintering process.
However, the PCI/SCC failure problem still waits for further progress to find a reliable, nonexpensive solution that also meets more recent requirements on fuel performance, e.g., imposed by fuel exposure to higher fuel burnup. So far, the reactor operators prefer to cope with the PCI/SCC performance problem by simply following cautious operating guide rules. This approach, however, may be costly because it restricts flexible power operation. Moreover, it is not reliable as a protection against PCI/SCC failure.
Currently the pure zirconium "liner" concept attracts commercial interest as a PCI/SCC remedy. It permits flexible power operation but is expensive in manufacturing and the zirconium liner may corrode excessively when exposed to high temperature water in failed fuel. Furthermore, that PCI/SCC remedy does not address the fission gas release (FGR) problem at medium or high burnup. On the contrary its use may even promote an excessive FGR, i.e., it offers the advantage of being insensitive to PCI/SCC failure on operation at attractive high power in the temperature range where, however, FGR takes place.
In heavy water reactors a graphite coating of the clad bore surface operates satisfactorily as a PCI/SCC remedy up to approximately 10 MWD/kg U. At the high exposures typical of LWRs the graphite coating, however, shows a very unreliable performance. Otherwise, the graphite coating technique is quite attractive as being quite simple and inexpensive.
The clad rib design of this prior art has been shown to raise the PCI/SCC failure resistance significantly and this is due to certain combinations of bore design parameters.
The present invention is based on the results and analysies of continued irradiation programs testing various modifications of the original internal clad rib design as disclosed in British Pat. No. 1 454 618. The aim has been to explore this fuel concept still further, as it has been realized that it has unique potential for solving also other types of fuel performance problems of more recent concern, in particular relating to fission gas release. As a result a still more effective, low-cost and technically simple design solution to the afore-mentioned PCI/SCC failure problem has emerged. Concurrently the same modified design offers solutions, unique to the rib cladding, to a number of additional, potentially life limiting performance problems, currently of special interest in the attempts to reach higher fuel burnup levels. Basically, all these performance problems being addressed emanate from one single crucial phenomenon, i.e., the excessive release of fission product gases and other volatile elements from the fuel pellets, operating at high temperature (&gt;1000.degree. C.), like xenon, krypton, iodine, cadmium, etc.
The fission gases (FG), that give rise to a variety of fuel performance problems, are known to be released significantly first above a certain fuel temperature level. This level decreases some hundred centigrades with increasing fuel burnup and may be exceeded during operation and thus more readily at higher burnup, e.g. on a power ramp. Iodine and cadmium are considered to be the active corrodents in the PCI/SCC failure mechanism. Xenon and krypton are heavy noble gases of very low thermal conductivity which on release contaminate the high conductivity helium filler gas in the fuel rod. As a result the thermal conduction within the fuel rod will be impaired, which causes the temperature to rise within the fuel pellets. If this temperature rise exceeds the critical temperature of FGR an additional amount of the contained FG in the pellet structure will be released. This causes an additional temperature rise and still more FGR etc., i.e. a "temperature feed back mechanism" comes into operation. The rise to high temperatures and the quantity of released fission gases become potentially life limiting. The higher fuel temperature, (resulting in higher stored fuel energy), and the release of large quantities of fission gases may present safety problems, e.g., in loss of coolant accidents (LOCA) and give rise to an impermissible, high internal gas pressure before "end-of-life".
The thermal feed back effect becomes occasionally operative already at medium burnup levels, particularly in BWR fuel rods, say at about 15-20 MWd/kg U, causing some fuel rods to show high FGR values, typically in excess of about 10% of the accumulated inventory. Only a limited number of such high FGR rods in the reactor core are generally tolerable form a LOCA point of view. The reasons for the onset of excessive FGR may be several: an inadvertent power transient, the presence of thermally unstable fuel pellets or the incorporation of too large initial pellet/clad gaps, etc. When once initiated the process tends to propagate spontaneously from pellet to pellet along the whole fuel column along with the successive degradation of the thermal conductivity by the migrating fission gases.
An important contributing factor to the thermal feed back effect is a very slow restoration of the gas conductivity at the position of FGR within the fuel rod. In particular, under conditions of fuel operation, when the helium fill gas in the top chamber (plenum) of the fuel column (see FIG. 1 of said British patent) has to migrate relatively long distances axially down the rod through narrow and tortuous paths within the cracked pellet column, the problem becomes accentuated. At high burnup levels, say beyond 30 to 40 MWd/kg U, where the pellet/clad gap is essentially closed the problem of slow axial gas mixing becomes a real concern. In fact, it has been experimentally observed that under closed gap conditions the axial gas migration is almost non-existent. It may take days or even weeks for the helium gas to migrate down to the lower part of the fuel rod and there successively restore the gas conductivity. The risk for initiation of a thermal feed back effect is obvious. The prospects for extending the fuel operation to higher burnup levels may thus be effectively restricted by the slow axial gas mixing process, should a critical amount of fission gases happen to be released during operation.
Another important experimental observation in this context is that in spite of the fact that the fuel rod of the conventional design operates with the pellet/clad gaps effectively closed and the pellet/clad contact pressure remains high, and consequently the interfacial heat transfer should be very good, the pellet average temperature rises significantly when fission gases are being introduced. The interpretation of this behaviour is, that the initial hot gap volume by the time has moved inwards into the pellet body (whose fragments have relocated outwardly) and there re-appears as a multitude of crack voids. This dispersed crack volume may amount to some 10% of the initial cold assembly gap space. When penetrated by fission gases, the narrow pellet cracks will act as thermal barriers and effectively impair the heat conduction through the pellet fragments.
In order to depress such inadvertent release of fission gases during power manoeuvering by means of fuel design and make it feasible to extend the fuel burnup target significantly, smaller diameter rods are currently becoming commercially introduced into the fuel assemblies, e.g., 9.times.9 rod assemblies instead of current 8.times.8 rod assemblies in BWRs. In doing so, the fuel operating temperature decreases as also the linear heat rating of the fuel rods. Consequently the margin against inadvertent release of fission gases becomes larger. Furthermore, the margin against PCI/SCC failure also improves. However, such rods, which are more expensive in fabrication, may suffer from another feature of heat conduction now appearing as a shorter fuel thermal time constant, i.e., such slimmer rods transfer the generated fissile power with a faster rate to the water coolant than larger diameter rods. This behaviour may effect the BWR core stability under certain modes of operation and present a safety issue. In this case the heat transfer is actually too efficient through the pellet/clad gap.